Streamlined best estimate plus uncertainty analysis of a GEN III plus BWR for a bottom drain line small break LOCA

Best Estimate Plus Uncertainty analysis has been used in the design, certification, and safety review of Nuclear Power Plants. To analyze the small break loss of coolant accident of a Generation III + Boiling Water Reactor, a streamlined BEPU method has been established and partially demonstrated to a Bottom Drain Line small break Loss Of Coolant Accident post-test analysis in this investigation. A RELAP5 input model was built for an integral test facility simulating a Generation III + Reactor with passive safety features. The Fast Fourier Transform Based Method was utilized to preliminarily evaluate the accuracy of the code simulation results. The DAKOTA toolkit was utilized to drive uncertainty quantification and sensitivity analysis. The simulation results generally reproduce the experimental phenomenon observed during experiments. The uncertainty bands of output pa-rameters were established to demonstrate the enclosing of experimental data and the safety margin according to the design limits.